The invention relates to a method of determining neutron spectra utilizing two neutron detectors with different sensitivity functions wherein the counting rates of the two detectors are integrated and the spectrum is determined therefrom and to a device for performing the method.
Such a method is known from J.Barthe et al., xe2x80x9cRadiation Protection Dosimetryxe2x80x9d, Vol. 70, Nos. 1-4, 59-66, (1997) Nuclear Technology Publishing.
For the measurement of the of neutron radiation doses mainly passive detection systems are utilized. These systems, however, are able to determine the doses sufficiently accurately only for limited energy ranges of the neutron radiation (for example, Albedo- or plastic trace detectors). In addition, these systems are analyzed only once per month so that an excessive radiation exposure is detected in some cases only after several months. It would be a large advance in personnal dosimetry detection of neutron radiation exposure if a real-time dose could be determined for an energy range covering thermal neutrons, that is, from the meV area to about 20 MeV without the need for workplacexe2x80x94specific calibration and correction factors. Most of the neutron fields, which normally occur (nuclear power plants, processing and transport of nuclear fuel, neutron therapy, etc.) have neutron energies in this range.
Personal dosimeters are used at present mainly in the field of photon detection. Examples of electronic personal dosimeters for photon radiation are: A personal dosimeter developed by Siemens-Plessey (EDPI, see for example, xe2x80x9cElectronic Dosimetry, 1/93 issue 1, Siemens Plessey Controls Ltd, 1933) for the representation of the new measurement value Hp (10) for photon radiation and for the detection and dosimetry of photon and electron radiation with small penetration ranges (skin dose), or the dosimeter of the company Rados (RAD-50, RAD-52, digital pocket dosimeters, data sheets of Rados Technology, Oy, Finland) which, like the Siemens dosimeter are based on Si diodes.
There are presently only few active personal dosimeters on the basis of silicon. Examples are the neutron dosimeters of the Japanese firm Aloka (Barthe, Bordy, Lahaye; Electronic Neutron Dosimeters: History and State of the Art, Radiation Protection Dosimetry, Vol. 70, Nos. 1-4, 59-66 (1997)). They are designed for the detection of thermal and fast neutrons (Model PDM-303, see data sheet ALOKA CO. LTD Tokyo Japan) or exclusively for thermal neutrons (model PDM-307, see data sheet). The systems of the firm Stephen, which are based on counter tubes are suitable to detect photon radiation and to perform a dosis evaluation (data sheet Stephen, Centronic D.C.A. Mini Instruments T.G.M.; Surrey England).
Neutron detection systems for determining information concerning the energy distribution of neutron radiation over the radiation protection energy range of 10 meV to about 10 MeV, are known so far only in the form of the xe2x80x9cBonner Kugelxe2x80x9d (Bonner spheres), which are based on the different absorption of the neutrons of each detection unit. (Bramblett H, Ewing, Bonner; A New type of Neutron Spectrometer; Nuclear Instruments and Methods 9, 1-12 (1980)). In an evaluation procedure, an experienced expert subsequently performs a so-called development in order to determine the neutron spectrum.
All the systems mentioned are so designed that they operate as occurrence counters without performing a radiation field analysis with the possibility of a spectrum determination. The Aloka dosimeter (PDM-303) was subjected to tests in order to determine the accuracy of a dose indication as a function of the neutron energy (for example CERN Report TIS-RP/TM/92-90 rev.(1992)). A result of the tests indicates a large deviation of the dose indication in comparison with the actual dose.
Of all these systems not a single one is suitable to represent, that is, to depict the neutron spectrum.
The only presently available system for neutron spectrometry covering the full energy range from thermal up to fastxe2x80x94that is the Bonner sphere system, is very expensive and involved since the detection units are very large and difficult to transport. Furthermore, they can be evaluated only by an experienced expert employing assumptions and information concerning the spectra measured. U.S. Pat. No. 5,572,028 discloses a detector system employing thermoluminescence detectors wherein dose values or energies are determined from the detector signals using artificial neural networks. The sensitivity functions of the various detectors are influenced in this case only by the use of various filters.
Furthermore, GB 1 014 682 discloses a method for determining neutron spectra wherein the energy spectrum is determined from the spectra measured by several threshold value detectors using a simple development procedure. This method however is limited to the energy ranges of 0.4 to maximally 3 MeV because of the type of threshold value detectors employed.
It is the object of the present invention to provide a method of the type described above, wherein, however, the neutron spectrum can be determined in real time and also to provide an apparatus for carrying out the method.
In a method and apparatus for determining neutron spectra using at least two neutron detectors which provide integral counting rates from which the spectrum of a neutron radiation can be approximated, and which consist each of a semiconductor diode, a converter layer, an inactive layer and an active layer. The various layers of each detector are different from those of the other neutron detector and so selected that the sensitivity functions of the two neutron detectors are different. An artificial neutral network is provided which is especially trained and to which the counting rates of the detectors are supplied to be processed for obtaining the neutron spectrum.
With the method according to the invention, wherein the dosis determination for neutron radiation is based on semiconductor detectors, an instant evaluation (real time dosimeter) is facilitated. The dosimeter consists of a multi-element system for the recording of signals and a corresponding evaluation algorithm based on the concept of artificial neural networks, wherein no preliminary information concerning the spectral energy distribution of the neutron radiation is needed.
In contrast to prior art measuring systems, the energy distribution of the neutron radiation is now determined in a first step and the measurement value is calculated in a second step. This has the advantage that, with the use of appropriate dose conversion factors, various dose values, for example, body and organ dose values can be provided which, in the case of excessive exposures, are important for contributory values for determining an appropriate exposure value.
A radiation sensor includes a semiconductor having a converter associated therewith. Such a radiation sensor can be considered as an element. Or, if the particles are registered by way of the recorded pulse level distribution, the energy deposition occuring in a sensor for a particular energy range can be considered to be the element. In the method, the incident neutron spectrum is correlated with the ratios of the recorded counter incidents of the particular elements since each element has an individual energy dependency of the recorded counter incidents of the incident neutron radiation. The evaluation algorithm provides in the first step the neutron spectrum with a suitable energy distribution and, in additional steps, measured dose values are determined which are present in the evaluation software in the form of tables and which are therefore easily adapted (for example, the personal dose Hp(10), effective dose).
The method presented herewith simplifies the measuring and evaluation procedure as it permits an immediate development after a certain counting period. Based on the retained information concerning the energy of the neutron radiation field, various dose measurement values can be determined which permit the formation of a single measurement value. In addition, a new measurement value may eventually be introduced by a simple adaptation of the evaluation software. For the detection systems used at the present time, the hardware would have to be changed in order to permit the introduction of a new measurement value.
The novelty of the invention resides in the determination of the spectral distribution of the neutron radiation (possibly in large energy intervals) since all the dose measurement values can be derived from the spectral distribution (taking also in consideration the incident angle with respect to the person exposed). The evaluation of the neutron radiation can be relatively easily modified to accommodate the introduction of a new measurement value (as it presently happens with the new ICRV measurement value by appropriate adaptation of the dose conversion factors in the evaluation algorithm.
It is therefore no longer necessary to adapt, that is to change, the detection hardware. It is also possible to indicate body and oxygen doses as it has been proposed by the radiation protection rules.
In summary, the properties and advantages of the invention are as follows:
Determination of the spectral information and the dose (for example, Hp(10)).
The use of a multi-element system (1) for receiving data: an element is defined herein by a variation of its output signal depending on the spectral make-up of the neutron field.
For the analysis of the input data, an artificial neural network (2) is utilized. This calculation method requires no setup information with respect to the type of radiation source (reactor, transport container, source, etc.). The system can be adapted to requirement changes simply by changing the software (not parameters).
The dose or, respectively, the dose spectrum is calculated by easily exchangeable dose conversion functions.
The system permits a time dependent measurement value determination. The measurement values can be indicated also during the measurement procedure (direct indication).
Since the system utilizes several elements for the data recording a dose indication is still possible if one or several elements are omitted.
Below, the invention will be described on the basis of two embodiments in connection with the accompanying drawings.